ML20151U882
| ML20151U882 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/08/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20151U880 | List: |
| References | |
| RTR-REGGD-01.101, RTR-REGGD-1.101 NUDOCS 9809110215 | |
| Download: ML20151U882 (12) | |
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION o,
WASHINGTON, D.C. 20066-0001
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SAFETY EVALUATION BY THE OFFICE OF NUCl FAR REACTOR REGULATION ON REVISED EMERGENCY ACTION LEVELS FOR GPU NUCIEAR. INC.
THREE MILE ISLAND NUCI FAR PLANT UNITS 1. and 2 DOCKET NOs. 50-289 and 50-320
1.0 INTRODUCTION
By letter dated December 2,1995, as supplemented by letters dated July 1,1997, May 14 and July 23,1998, GPU Nuclear, Inc. (GPU, the licensee) proposed changes to the Three Mile Island Nuclear Station (TMI) Units 1 and 2 emergency action levels (EALs). Specifically, the licensee provided Appendix A to the GPU Nuclear Power Radiological Emergency Plan, which included the initiating condition for event classes, applicable plant conditions, and EALs; Emergency Plan implementing Procedure (EPIP-TMI.01, Revision 5); " Emergency Classification Procedure;" and technical basis documentation. These collectively described how the proposed EALs incorporated the guidance in NUMARC/NESP-007, " Methodology for Development of Emergency Action Levels," Rev,2, January 1992. NUMARC/NESP-007 was endorsed by the NRC in Regulatory Guide 1.101, " Emergency Planning and Preparedness for Nuclear Power Reactors," Rev. 3, August 1992, as an acceptable method by which licensees may develop site-specific emergency classification schemes for meeting the planning standard of 10 CFR 50.47(b)(4) and the requirements of Appendix E to 10 CFR 50.
The staff completed its initial review of the proposed revision to the TMI EAL scheme dated December 2,1995, and requested additionalinformation on August 22,1996. A public meeting was conducted on September 24,1996, to discuss the licensee's responses to the request for additional information. At that meeting, the licensee agreed to modify its proposed revision to include the responses to the request for additional information and submitted a revised EAL scheme on July 1,1997. Following telephone conference calls with the licensee in the q
November 1997 to January 1998 time frame, the licensee modified the July 1,1997, proposed revision and provided a revised EAL scheme on May 14,1998. Following telephone conference calls in the June / July 1998 time frame, the licensee modified the May 14,1998, proposed revision and provided a revised EAL scheme on July 23,1998.
2.0 REGULATORY BACKGROUND Title 10 of the Code of Federal Regulations (CFR) Part 50.47(b)(4) states, in part: "A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee... "
Appendix E, Subsection IV.B states, in part: "... These emergency action levels shall be discussed and agreed on by the applicant and State and local governmental authorities add anoroved by the NRC... " (Emphasis added) 9909110215 980908 f
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Appendix E, Subsection IV.C states, in part: " Emergency action levels (based not only on onsite and offsite radiation monitoring information but also on readings from a number of sensors that indicate a potential emergency, such as pressure in containment and response of the Emergency Core Cooling System) for notification of offsite agencies shall be described...
The emergency classes defined shall include (1) notification of unusual events, (2) alert, (3) site area emergency, and (4) general emergency... "
Regulatory Guide (RG) 1.101, Rev. 3, " Emergency Planning and Preparedness for Nuclear
. Power Reactors," endorsed NUMARC/NESP-007, Rev. 2, " Methodology for Development of Emergency Action Levels," as an acceptable method for licensees to meet the requirements of 10 CFR 50.47(b)(4) and Appendix E to 10 CFR 50.
3.0 EVALUATION The licensee provided Event Classification Matrices in EPIP-TMI.01, " Emergency Classification and Basis," Revision 5, which organized 32 initiating Conditions (ICs) into 7 EAL tables. Each table contained the operating condition applicability, the EAL statement (s), the EAL number, and the emergency class. Following each EAL table was a summary of the basis for the EAL that cited specific TMl procedures, equipment and instrument numbers and identified the corresponding NUMARC/NESP-007 Example EAL(s). The EAL tables are organized as follows:
Category 1.0 Radiological Controls i
Category 2.0 Fission Product Barrier i
Category 3.0 Electrical Category 4.0 Instrumentation, Actuation and Tech Specs Category 5.0 Natural Phenomena i.
Category 6.0 Man-Made Phenomena Category 7.0 Security.
Category 8.0 Judgement i
Each category is divided into one or more subcategories, titled with a brief statement of the IC.
For example: Category 1.0, Radiological Controls, contained the following subcategories:
1.1 Airbome Effluent Monitoring 1.2 Radioactive Material Control 1.3 Liquid Effluent Monitoring 1.4 Spent Fuel Pool 1.5 Reactor Cavity 1.6 Fuel Clad Degradation The staff relied upon the guidance in NUMARC/NESP-007 as the basis for its review of the TMI EAL changes. Most of the proposed EALs conform closely to the guidance. However, several of the licensee's proposed EALs depart from the example EALs in NUMARC/NESP-007. As 4
discussed below, the staff reviewed the licensee's justification for these deviations and found i,
them to be acceptable.
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3 3.1 DEVIATIONS FROM NUMARC/NESP-007
- 1. ICs were not included for some EALs The NUMARC/NESP-007 guidance specified ICs for grouping EALs. The ICs serve two purposes (1) to assist the emergency director in understanding the plant condition of concern (which is indicated by exceeding the EAL th'reshold; this also assists in the use of judgement in l
classifying events), and (2) conveying the plant condition of concem to offsite agencies (rather than providing the EAL threshold which was exceeded). The licensee did not explicitly group the EALs under ICs. However, most EALs contain introductory phrases which serve as ICs; L
and, for those which did not, the EAL itself serves the purpose of IC by the manner in which it is worded. Therefore, this deviation is acceptable.
NUMARC/NESP-007 IC AU1 states:
AU1 Any unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds Two Times the Radiological Technical Specifications for 60 Minutes or Longer.
The NUMARC/NESP-007 guidance for AU1 includes example EALs based on effluent monitor l
readings, sample results, perimeter radiation monitoring results and dose calculations. The example EAL for the perimeter monitoring sistem is as follows:
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- 3. Valid reading on perimeter radiation monitoring system greater than.1 mR/hr above
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normal background for 60 minutes [for sites having telemetered perimeter monitors).
J The licensee has an offsite monitoring (telemetric perimeter monitor) system. However, this l
NUMARC/NESP-007 example EAL was not included in its EAL scheme because the Reuter Stokes monitors used in this system are located at varying distances from the site - some inside
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L the exclusion area boundary and others up to several miles away. Monitors located beyond the l
exclusion area boundary are not under the direct control of the licensee; they can be affected i
by something other than an incident at TMI. Since this portion of the system is not under the r
direct control of the licensee and could provide erroneous information for classifying emergencies, and since the licensee included the other example EALs in this IC, the omission of this EAL is acceptable. This deviation is also applicable to NUMARC/NESP-007 ICs AA1, AS1, and AG1. These are also acceptable.
- 3. Omitted EALs based upon Critical Safety Function Status (CSFS)
Table 4 of NUMARC/NESP-007, provides example thresholds for the or potential loss of barriers for PWR Emergency Action Level Fission Product Barriers some of which are related to the CSFS. The licensee did not include EALs for the loss (and potential loss) of fission product barriers based upon CSFS. The licensee's emergency operating procedures (EOPs) do not include steps for monitoring Critical Safety Functions. The licensee included other EALs for 1
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4 indication of the loss or potential loss of fission product barriers. In that the licensee does not i
include the use of CSFS indicators in the EOPs, and has included other acceptable EALs for l
the loss or potential loss of the fission product barriers, this deviation is acceptable.
4.
Omitted EAL - FC4, for the Potential Loss of Fuel Clad Barrier Based Upon Reactor Vessel Level (RVL)
The NUMARC/NESP-007 EAL for the potential loss of the fuel clad barrier based on RVL is:
Reactor Vessel Water Level less than (site-specific) value
- The licensee's EAL scheme did not include this specific EAL, but did include an EAL for the potentialloss of the Fuel Clad Barrier based upon an indication of superheat, as follows:
> 25 'F Superheat The licensee indicated that the potential loss of the Fuel Clad Barrier can be determined by calculating a superheat value using the incore thermocouples and reactor coolant system (RCS) pressure. In order for the thermocouples to indicate superheat, the top of the core must
> be uncovered. This indirect measurement of reactor coolant inventory is an improvement over the Reactor Coolant inventory Trending system, which is not qualified and only trends reactor vessel level to the bottom of the hot leg piping. The staff agrees with the licensee's rationale for using an indication of superheat in place of a reactor vessel water levelindication. Therefore, l
this deviation is acceptable.
5.
Modified EALs for the Potential Loss and Loss of Fuel Clad Barrier and the Potential Loss of Containment Barrier Based Upon Core Exit Thermocouple Readings The NUMARC/NESP-007 EAL for the loss and potential loss of the fuel clad barrier based on core exit thermocouples (CETs) readings are:
Potential Loss of Fuel Clad:
Core Exit Thermocouple Readings f,reater than (site-specific) degree F l
(The basis for this EAL specifies that the reading should correspond to a loss of subcooling of the coolant... usually about 700 to 900 degrees F)
Loss of Fuel Clad:
Core Exit Thermocouple Readings greater than (site-specific) degree F (The basis for this EAL specifies that the reading should correspond to significant superheating i
of the coolant... usually about 1200 degrees F) r
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l Potential Loss of Containment:
Core exit thermocouples in excess of 1200 degree F and restoration procedures not effective within 15 minutes; or, core exit thermocouples in excess of 700 degree F with reactor vessel water level below top of active fuel and restoration procedures not effective within 15 minutes l
(The basis for the use of elevated core exit thermocouple reading's as an indication of a potential loss of containment is that these conditions represent an imminer.t melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure.)
The TMI EAL scheme includes the following EALs for these conditions:
Potential Loss of Fuel Clad:
> 25'F Super Heat Loss of Fuel Clad:
Tem 21400 'F Potential Loss of Containment:
Tem 21800 'F l
L The licensee chose to use RCS at 25 'F superheat as an indication of the potentialloss of the fuel clad instead of a core exit thermocouple reading. The licensee states that at this point the RCS has clearly moved from a subcooled to a superheated condition. With the system being superheated, the heat transfer mechanics are not as efficient and could lead to a rapid rise in cladding temperatures.
The licensee uses indication of projected clad temperature rather than core exit thermocouple readings for the loss of fuel clad and the potential loss of the containment barrier. The projected condition is a function of thermocouple indications and system pressure. The j
licenses states that this is a good correlation with the expected degradation of the fuel. The i
licensee chose to use an indication of Tem temperature of 1400 'F as an indication of the loss of fuel clad. The licensee states that a Tem temperature of 1800 'F is more conservative than a core exit thermocouple temperature of 1200 'F and that it selected 1400 *F for this setpoint because the duration to go from 1400 'F to 1800 'F is extremely short.
The licensee elected to use Tem of 1800 'F as an indication of the potential loss of containment. However, the licensee elected not to provide 15 minutes before classifying the containment as potentially lost when Tem is 1800 *F because it does not anticipate altering its conclusion in the 15 minute window.
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The staff concludes that the licensee's EALs are appropriate indicators for the loss or potential loss of fission product barriers. Therefore, these deviations are acceptable.
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- 6. Omitted EAL for loss of RCS based on Steam Generator Tube Rupture (SGTR) with a prolonged release The NUMARC EAL for the loss of RCS based on a SGTR is:
(Site-specific) indication that an SG is ru'ptured and has a non-isolable secondary line break <OR>
(site-specific) indication that an SG is ruptured and a prolonged release of contaminated secondary coolant is occurring from the affected SG to the environment The intent of this EAL is to classify events where there is a prolonged release from an SGTR as a Site Area Emerger,ty. The licensee has included two EALs which, if both are exceeded, result in a Site Area Emergency classification, i.e.:
Primary to Secondary Leakage > 160 gpm, and Total OTSG leak >1 gpm TS to atmosphere These EALs serve the intent of the omitted NUMARC/NESP-007 EAL. Therefore, this deviation from the NUMARC/NESP-007 guidance is acceptable.
- 3. Report of any visible structural damage on any of the following plant structures:
Reactor Building Intake Building Ultimate Heat Sink Refueling Water Storage Tank Diesel Generator Building Turbine Building Condensate Storage Tank Control Room The licensee indicated it does not use a specific EAL for each of the plant structures in this example but provided an attemative set of EALs which included a listing of buildings to be considered [EALs A4.2 (High Winds), A4.3 (Tornado) and A4.4 (Earthquake)). The following buildings were identified in these EALs:
Reactor Building, intake Building, intermediate Building, Control Tower,
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Aux & Fuel Handling Building, and Diesel Generator Building.
l in that the licensee has a specific listing of buildings identified in other EALs which meet the intent of the NUMARC example EAL, this deviation is acceptable.
- 8. Omitted List of Areas in Fire EAL HA2.1 '
NUMARC/NESP-007 EAL HA2.1 is:
l a.
Fire or explosion in any of the following (site-specific) areas:
(site-specific) list AND b.
Affected system parameter indications show degraded performance or plant personnel report visible damage to permanent structures or equipment within the specified area.
1 The corresponding TMI EAL is.
A5.1 FIRE 1.
Fire affects operability of 1 safety system train
<OR>
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2.
Fire inside Protected Area requires off-site assistance The licensee indicates it does not use a specific EAL for each of the plant structures in this example but covers them in EALs for Fire US.1, A5.1 and Explosion US.4, A5.4. Thia deviation is acceptable.
- 9. Omitted Turbine Missile EAL HA1.6 NUMARC/NESP-007 EAL HA1.6 is:
Turbine failure generates missiles result in any visible structural damage to or penetration of any of the following plant areas (site-specific list)
The licensee stated that it did not include this EAL because a turbine missile cannot reach the components vital to safe shutdown at TMI based on the design of the buildings housing the equipment and equipment location. The staff concludes that the licensee's rationale for not including this EAL is appropriate. Therefore, this deviation is acceptable.
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8 3.2 SITE-SPECIFIC SETPOINTS 1.
Setpoint for Potential Loss of Reactor Coolant System (RCS) Barrier Based Upon RCS Leak Rate The NUMARC/NESP-007 EAL for the potentialloss of the RCS barrier based on RCS leakage is:
Unisolable leak exceeding the capacity of one charging pump in the normal charging mode The NUMARC/NESP-007 basis for this EAL states:
The " Potential Loss" is based on the inability to maintain normal liquid inventory within the RCS -
by normal operation of the Chemical and Volume Control System which is considered as one centrifugal charging pump discharging to the charging header.
RCS leakage > 160 gpm in its December 2,1995, EAL submittal, the licensee used 120 gpm as the leak rate for this EAL and justified using this setpoint as follow:
"The 120 gpm value exceeds t apacity of a single make up pump feeding the Reactor Coolant System througn the normal make up valve (MU-V-17). This is consistent with the NUMARC/NESP-007 guidance."
in its revised submittal dated May 14,1998, the licensee modified this setpoint to 160 gpm stating that:
"The new value references the high flow alarm condition (~160 gpm). This indicator is likely to be determined easier and quicker than actually calculating the leak value. Hence, this will be a more timely evaluation of the barrier. The value chosen satisfies the NESP-007 criteria where the potentialloss is based 3
on the inability to maintain normal inventory control."
This site-specific setpoint is acceptable.
- 2. Potential Loss of RCS Integrity Indicated by Exceeding Heatup Limits One of the NUMARC/NESP-007 EALs for the potentialloss of RCS is:
Critical Safety Function RCS Intearity - Red The licensee does not utilize critical safety functions in its EAL scheme but addressed the condition indicated by a "RCS Integrity - Red" by including the following site-specific EAL:
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9 Exceed oressure/temoerature limits of TS HU/CD Curve This site-specific EAL is acceptable.
- 3. Decay Heat Removal EAL NUMARC/NESP-007 IC SS5 is:
SS5 Luss of Water Level in the Reactor Vessel That Has or Will Uncover Fuel in the Reactor Vessel.
The NUMARC/NESP-007 EALs unoer this IC are:
Loss of all decay Heat removal cooling as determined by site-specific procedure AND (site-specific) indicator that the core is or will be uncovered The licensee's EAL scheme includes the following Site Area Emergency EAL corresponding to this EAL:
< OR >
2b) Core exit temperature indicates 225 *Superheat The site-specific indications used in this EAL meet the intent of the NUMARC/NESP-007 guidance and are acceptable.
3.3 SITE-SPECIFIC ADDITIONS
- 1. TMI EAL Loss of Fuel Clad Barrier - Letdown Line Radiation Monitor The licensee's EAL scheme includes the following indication of the loss of the fuel clad barrier:
3 Letdown Line > 15 R/hr The licensee states that this is indicative of 5% clad damage and that the letdown line reading (which is taken at the letdown monitor) provides a quick approach to ascertain this minimum level without the delay associated with a post-accident sample. The staff agrees with this i
4 10 anticipatory approach given the inherent delay in obtaining a post-accident sampling system sample. Therefore, this additional EAL is acceptable.
' 2.
TMI EAL Potential Loss or Loss of RCS Barrier - Cycling or Stuck Open PORV or RCS Code Safety Valves The licensee's EAL scheme includes the following EAL for the Loss of the RCS boundary:
l Stuck open PORV OR RCS Code Safety Valve and the following EAL for the Potential Loss of the RCS Boundary:
Cycling PORV OR RCS Code Safety Valves The NUMARC/NESP-007 guidance did not specify these EALs but does allow for site-specific indications of the loss or potential loss of fission product boundaries. The cycling of a PORV is indicative of a potentialloss of RCS boundary and a non-isolated, stuck open PORV or a stuck open Code safety valve creates a non-isolable breach of the RCS and therefore constitutes a loss of the RCS barrier. Therefore, these additional site-specific EALs are acceptable.
- 3. TMI EAL for the Loss of the Containment Boundary The licensee added the following EAL to its classification scheme for the loss of the containment boundary based upon containment pressure:
RB Press 2100 psig The NUMARC/NESP-007 guidance did not specify a loss of containment boundary EAL based upon containment (or rea' tor building) pressure. The licensee provided the following c
justification for adding this EAL.
"An analysis was performed to verify integrity of the containment as a barrier to the release of fission products. This showed that, mathematically, the building would be '
intact at up to three times the design pressure or 150 psig. The margin of safety would be greatly decreased at that point. The calculations showed that cracking could be expected at 120 psig. Therefore, a conservative value of 100 psig was assumed to be the point where the containment barrier was lost."
The staff agrees with the licensee's rationale for including this EAL as an indication of the loss of the containment barrier. Therefore, this site-specific EAL is acceptable.
The licensee's EAL scheme includes the following EAL for the loss of containment based upon l
plant exhaust radiation readings:
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RM-A-8GH E 200 cpm (Gas High Range)
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11 The NUMARC/NESP-007 guidance did not specify a loss of containment boundary EAL based upon plant exhaust radiation monitor readings. The licensee provided the following justification for adding this EAL.
"This is indicative of a 120 gpm leak with the RCS activity 2 Tech Spec, assuming leakage in the Auxiliary building that cannot be isolated. This provides for fission products to be outside the containment barrier (bypassed) and can be considered as lost."
The staff agrees that the licensee's EAL is an appropriate indication of the loss of the containment barrier when an RCS leak is occurring. Therefore, this site-specific EAL is acceptable.
- 5. TMl EAL Unusual Event for a SGTR >10 gpm to the Condenser The licensee's EAL scheme includes the following Unusual Event EAL:
Total OTSG 210 gpm to the condenser The NUMARC/NESP-007 guidance did not specify this specific EAL but does include an EAL based _upon RCS pressure boundary leakage. The licensee's EAL is equivalent to a 10 gpm i
pressure boundary leak. The licensee also includes the following EAL as a indication of the loss of containment boundary:
Total OTSG 21 gpm TS to atmosphere This EAL is also classified as an U.nusual Event if no other fission product barriers have been lost. The steam generator leak to the condenser is not considered a release to the atmosphere and therefore is not considered a loss of containment under the NUMARC/NESP-007 guidance.
Since this EAL is consistent with the NUMARC/NESP-007 guidance for an Unusual Event iNL for based upon RCS leakage, this site-specific EAL is acceptable 4.0 STATE AND LOCAL GOVERNMENTAL ORGANIZATION REVIEW OF TMl EALs Appendix E, Subsection IV.B states, in part: "... These emergency action levels shall be discussed and aareed on by the aoolicant and State and locel aovemmental authorities and approved by the NRC... " (Emphasis added)
In its December 5,1995, submittal, the licensee indicated that the Commonwealth of Pennsylvania, Pennsylvania Emergency Management Agency (PEMA) had conducted a review of the proposed changes in the EALs for the Three Mile Island Nuclear Plant. The licensee stated that comments from PEMA and the Commonwealth of Pennsylvania Bureau of Radiation Protection were incorporated in the TMI EALs submitted on December 5,1995. However, because revisions have been made since that time, the revisions need to be discussed with the Commonwealth of Pennsylvania and local governmental authorities. In its July 23,1998, letter, the licensee stated: "This proposed EAL revision will be discussed and agreed upon with State and local governmental authorities prior to implementation." Consequently, prior to
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implementation of this EAL scheme, the licensee will have to have the Commonwealth of Pennsylvania and loca! govemmental agreement.
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CONCLUSION The proposed EAL changes for TMI are consistent with the guidance in NUMARC/NESP-007, with variations as identified and accepted in this review, and, therefore meet the requirements of 10 CFR 50.47(b)(4) and Appendix E to 10 CFR Part 50.
Prior to implementation of the TMI EAL scheme, the licensee will have to obtain the Commonwealth of Pennsylvania's and local govemmental organizations' agreement in accordance with the requirements in Appendix E to 10 CFR Part 50.
Principal Contributors: E. Fox, Jr.
S. Roudier J. O'Brien Date: September 8,1998 t
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